NRC-15-0068, License Renewal Application 2015 Annual Update

License Renewal Application 2015 Annual Update

ML15180A327
Person / Time
Site:	Fermi
Issue date:	06/26/2015
From:	Kaminskas V DTE Energy
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NRC-15-0068
Download: ML15180A327 (46)

Text

{{#Wiki_filter:Vito A. Kaminskas Site Vice President DTIE Energy Company 6400 N. Dixie Highway, Newport, MI 48166 Tel: 734.586.6515 Fax: 734.586.4172 Email: kaminskasvidteenergy.com i DTE Energy 10 CFR 54 June 26, 2015 NRC-15-0068 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington D C 20555-0001

References:

1) Fermi 2 NRC Docket No. 50-341 NRC License No. NPF-43

2) DTE Electric Company Letter to NRC, "Fermi 2 License Renewal Application," NRC-14-0028, dated April 24, 2014 (ML14121A554)

Subject:

Fermi 2 License Renewal Application 2015 Annual Update In Reference 2, DTE Electric Company (DTE) submitted the License Renewal Application (LRA) for Fermi 2. The purpose of this letter is to provide the NRC an update to the Fermi 2 LRA in accordance with 10 CFR 54.21(b). During NRC review of the Fermi 2 LRA, DTE is required by 10 CFR 54.21(b) to provide an amendment to the Fermi 2 LRA that identifies changes to the Fermi 2 current licensing basis (CLB) that materially affects the contents of the Fermi 2 LRA each year and at least three months before scheduled completion of the NRC review. DTE has completed the annual CLB review and is providing the changes to the LRA required by this review in Enclosure 1. In addition, Enclosure 1 of this letter includes additional LRA revisions that have been identified but are not related to changes to the Fermi 2 CLB. No new commitments are being made in this submittal. Should you have any questions or require additional information, please contact Lynne Goodman at 734-586-1205.

USNRC NRC-15-0068 Page 2 I declare under penalty of perjury that the foregoing is true and correct. Executed on June 26, 2015 Vito A. Kaminskas Site Vice President Nuclear Generation

Enclosures:

1. Fermi 2 License Renewal Application Revisions for 2015 Annual Update cc: NRC Project Manager NRC License Renewal Project Manager NRC Resident Office Reactor Projects Chief, Branch 5, Region III Regional Administrator, Region III Michigan Public Service Commission, Regulated Energy Division (kindschl@michigan.gov) to NRC-15-0068 Fermi 2 NRC Docket No. 50-341 Operating License No. NPF-43 Fermi 2 License Renewal Application Revisions for 2015 Annual Update to NRC-15-0068 Page 1 Current Licensing Basis Changes In accordance with 10 CFR 54.21(b), DTE has reviewed changes to the Fermi 2 current licensing basis (CLB) to determine if changes impact the Fermi 2 License Renewal Application (LRA).

The results of this review that impact the LRA are discussed below. A. Nonsafety-related flexible hoses in the reactor recirculation system (B31) were replaced. The material was previously stainless steel. The new material is nickel alloy. Nonsafety-related components in the B31 system were reviewed in LRA Table 3.1.2-4-2. There are existing line items in LRA Table 3.1.2-4-2 for stainless steel flex connections and these are retained since not all of the nonsafety-related flexible hoses in the reactor recirculation system were replaced. LRA Table 3.1.2-4-2 is revised to add line items for the new nickel alloy flex connections. In addition, LRA Section 3.1.2.1.4 is revised to add nickel alloy to the list of materials and LRA Table 3.3.1 (Item 3.3.1-95) is revised to include nickel alloy in the description of materials. B. Strainers and strainer housings in the emergency equipment cooling water system (P44) were replaced. The strainer and strainer housing materials were previously copper alloy and carbon steel, respectively. The new strainer and strainer housing material is stainless steel. Although these components were part of the emergency equipment cooling water system, they were reviewed under the compressed air system in LRA Table 3.3.2-6. There are existing line items in LRA Table 3.3.2-6 for stainless steel strainer housings, so no new line items were needed for the housings. LRA Table 3.3.2-6 is revised to change line items for copper alloy strainers to stainless steel and to delete line items for carbon steel strainer housings. C. Tubing in the emergency diesel generator system (R30) was replaced. The tubing material was previously copper alloy. The new material is stainless steel. The emergency diesel generator system is reviewed in LRA Table 3.3.2-10. There are existing line items in LRA Table 3.3.2-10 for copper alloy and these are retained since not all of the tubing was replaced. LRA Table 3.3.2-10 does contain existing line items for stainless steel tubing, but not for the internal environment of this new tubing. LRA Table 3.3.2-10 is revised to add a line item for stainless steel tubing in a treated water environment (not greater than 140°F). D. Tubing in the fuel oil lines of the emergency diesel generator system (R30) was replaced. The tubing material was previously carbon steel. The new material is stainless steel. The fuel oil portion of the emergency diesel generator system is reviewed in LRA Table 3.3.2-15. There are existing line items in LRA Table 3.3.2-15 for stainless steel tubing, so no new line items were needed for the new material. However, there are no existing line items for carbon steel tubing (it previously was covered by line items for "piping"). Since not all of the carbon steel tubing was replaced and to ensure consistency throughout the LRA with respect to the use of "piping" vs. "tubing", LRA Table 3.3.2-15 is revised to add line items for carbon steel tubing. to NRC-15-0068 Page 2 E. Nonsafety-related valves in the heater drains system (N22) were modified to add nickel alloy (Inconel) cladding to the internal surfaces. The valves were previously carbon steel with no internal cladding. The nonsafety-related components in the N22 system were reviewed in LRA Table 3.4.2-3-4. There are existing line items in LRA Table 3.4.2-3-4 for carbon steel valve bodies. LRA Table 3.4.2-3-4 is revised to add line item for the new carbon steel valve bodies with nickel alloy internal cladding. F. The nonsafety-related steam jet air ejectors (SJAEs) in the condenser and auxiliaries system (N61) were modified. The SJAEs were previously identified in the LRA as stainless steel ejectors. Review of the modification identified that the stainless steel components are internal to the ejector and that the material that performs the (a)(2) pressure boundary function is carbon steel. The nonsafety-related components in the N61 system were reviewed in LRA Table 3.4.2-3-6. The existing line items in LRA Table 3.4.2-3-6 for stainless steel SJAEs are deleted. LRA Table 3.4.2-3-6 is revised to add line items for the carbon steel SJAEs. In addition, a line item for an internal environment of steam was added since the SJAEs may be exposed to both steam and treated water. G. Relief valves in the lube oil filters in the emergency diesel generator system (R30) were identified for replacement. The valve body material will be changed from carbon steel to aluminum but the change has not yet been implemented. However, from the review it was determined that the carbon steel material is exposed to both an internal and external lube oil environment. There is an existing line item in LRA Table 3.3.2-10 for carbon steel valve bodies with an internal lube oil environment. LRA Table 3.3.2-10 is revised to add a line item for carbon steel valve bodies with an external lube oil environment. H. Metal fatigue calculations were revised and an environmentally assisted fatigue (EAF) screening was performed. LRA Sections 4.3, 4.6, 4.7, 4.8, and A.2.2 are revised to show the results of the fatigue calculations and reflect the EAF screening. More detailed calculations were performed for the locations where the projected cumulative usage factor did not meet acceptance criteria at the time of the LRA submittal, and the revised calculations show the criteria is met, excluding the impact of EAF. The EAF screening has identified some locations where additional action will be needed, e.g. more detailed analysis or stress based monitoring. LRA Sections 4.3 and A.2.2.3 are revised to indicate that the Fatigue Monitoring Program (LRA Section B.1.17) includes stress-based or cycle-based fatigue monitoring. Although the EAF screening has been performed, the commitment to develop EAF usage calculations (LRA Table A.4, Item 12.b) remains, since additional action is needed based on the EAF screening results. I. Procedures associated with the Water Chemistry Control - BWR Program were revised to use the 2015 revision of BWRVIP-190 (EPRI Report 3002002623) rather than 2008 revision (EPRI Report 1016579). The Water Chemistry Control - BWR Program descriptions in LRA Sections A.1.43 and B.1.43 are revised to update the EPRI report number. NUREG-1801 Section XI.M2 also references the 2008 revision (EPRI Report to NRC-15-0068 Page 3 1016579) in the program description. The use of the 2015 revision of BWRVIP-190 can be considered an exception to NUREG-1801. LRA Section B.1.43 is revised to include discussion of this exception to NUREG-1801. LRA Table B-3 is also revised to indicate that the Water Chemistry Control - BWR Program has an exception. In addition, the tables in LRA Section 3.0 include a "Notes" column. Generic notes A and B are identical except that note A is used for programs that are consistent with NUREG-1801 and note B is used for programs that have exceptions to NUREG-1801. The same is true for generic notes C and D. Due to the new exception discussed above, any line item in a table in LRA Section 3.0 that references the Water Chemistry Control - BWR Program and uses note A is revised to use note B. Similarly, LRA Section 3.0 table line items that reference the Water Chemistry Control - BWR Program and use note C are revised to use note D. Since this change affects multiple line items in multiples tables in LRA Section 3.0 and the change is administrative in nature, actual markups to the existing LRA tables are not provided in this letter. Administrative Changes In addition to the CLB changes, DTE has also identified some other non-material changes to the LRA. These changes are administrative changes to ensure consistency between the LRA and previously submitted RAI responses. These additional items that impact the LRA are discussed below. J. In Enclosure 2 to DTE letter NRC-15-0030, dated March 19, 2015, DTE revised LRA Table 3.2.2-8-1 to add a line item for piping. The line item credits the Buried and Underground Piping aging management program. For consistency with that change, LRA Section 3.2.2.1.8 is revised to add Buried and Underground Piping to the list of aging management programs for the miscellaneous ESF systems in scope for 10 CFR 54.4(a)(2). K. In the response to RAIs B.l.19-2a and B.1.19-8a in DTE letter NRC-15-0031, dated April 10, 2015, DTE added item "q" and modified item "p" in the Fire Water System Program in LRA Table A.4. However, the new item should have been labeled "r" and the item being modified should have been labeled "q". LRA Table A.4 is revised to clarify the item sequence. No changes to the text are being made, only to the sequence letters. L. In the response to RAI B.1.17-2 in DTE letter NRC-15-0005, dated January 20, 2015, DTE added enhancements to the Fatigue Monitoring Program in LRA Section B.1.17. In that letter, the two new enhancements were shown as applicable to program elements 1 through 7. However, the first new enhancement was only applicable to program elements 1 through 6 and the second new enhancement was only applicable to program element 7 (i.e. the enhancements should have been separated by a table line). LRA Section B.1.17 is revised to clarify which program elements apply to the enhancements. No changes to the text are being made, only to the format of the table. to NRC-15-0068 Page 4 M. As discussed previously, the tables in LRA Section 3.0 include a "Notes" column and generic notes A and B are identical except that note A is used for programs that are consistent with NUREG-1801 and note B is used for programs that have exceptions to NUREG-1801. The same is true for generic notes C and D. In previous DTE letters, exceptions to NUREG-1801 were added or deleted from certain aging management programs. The line items in LRA Section 3.0 tables that reference these aging management programs and use a generic note A, B, C, or D are revised to use the appropriate note as indicated in the list below. Since these changes affect multiple line items in multiples tables in LRA Section 3.0 and the changes are administrative in nature, actual markups to the existing LRA tables are not provided in this letter. " DTE letter NRC-15-0056 dated May 19, 2015 revised the BWR Feedwater Nozzle Program (LRA Section B.1.6) to add an exception to NUREG-1801. For Section 3.0 table line items that reference this program, all use of generic note A associated with this program is revised to note B and all use of generic note C is revised to note D. DTE letter NRC-15-0001 dated January 5, 2015 revised the BWR Stress Corrosion Cracking Program (LRA Section B.1.8) to add an exception to NUREG-1801. For Section 3.0 table line items that reference this program, all use of generic note A associated with this program is revised to note B and all use of generic note C is revised to note D. " DTE letter NRC-14-0051 dated July 30, 2014 revised the Fire Water System Program (LRA Section B.1.19) to add exceptions to NUREG-1801. For Section 3.0 table line items that reference this program, all use of generic note A associated with this program is revised to note B and all use of generic note C is revised to note D. DTE letter NRC-15-0001 dated January 5, 2015 revised the Reactor Vessel Surveillance Program (LRA Section B.1.38) to delete the exception to NUREG-1801. For Section 3.0 table line items that reference this program, all use of generic note B associated with this program is revised to note A and all use of generic note D is revised to note C. " DTE letter NRC-15-0030 dated March 19, 2015 revised the Water Chemistry Control - Closed Treated Water Systems Program (LRA Section B.1.44) to add an exception to NUREG-1801. For Section 3.0 table line items that reference this program, all use of generic note A associated with this program is revised to note B and all use of generic note C is revised to note D. N. LRA Section A.3 is a list of references for Appendix A of the LRA. In the original LRA submittal, the first reference (A.3-1) was a placeholder for the LRA submittal. LRA Section A.3 is revised to replace the placeholder reference for the LRA submittal with the actual reference. to NRC-15-0068 Page 5 LRA Revisions The LRA revisions associated with the changes discussed above (both the CLB-related and administrative changes) are shown on the following pages. Additions are shown in underline and deletions are shown in strike-through. Note that previous changes made to these same LRA sections made in previous letters are not shown in underline or strike-through such that only the new changes due to the items below are shown as revisions. Since a single change may result in an impact to multiple locations in the LRA, the LRA revisions are labeled with a letter in the right margin that cross-references the change to the list above. The LRA revisions are provided in the order that they would appear in the LRA. to NRC-15-0068 Page 6 3.1.2.1.4 Miscellaneous RCS Systems in Scope for 10 CFR 54.4(a)(2) Materials Nonsafety-related components affecting safety-related systems are constructed of the following materials. Carbon steel Copper alloy

• Copper alloy >15% zinc or >8% aluminum

" Glass Nickel alloy l A Stainless steel to NRC-15-0068 Page 7 Table 3.1.2-4-2 Reactor Recirculation System Nonsafety-Related Components Affecting Safety-Related Systems Summary of Aging Management Evaluation Table 3.1.2-4-2: Reactor Recirculation System, Nonsafety-Related Components Affecting Safety-Related Systems Aging Effect Aging Component Intended Requiring Management NUREG-Table 1 Type Function Material Environment Management Program 1801 Item Item Notes Filter housing Pressure Carbon Lube oil (int) Loss of Oil Analysis V.D2.EP-77 3.2.1-49 C, 103 boundary steel material Flex Pressure Nickel Air-indoor None None IV.E.RP-03 3.1.1-A connection boundary alloy (ext) 106 Flex Pressure Nickel Waste water Cracking - TLAA-metal H connection boundary ajloy (int fatigue fatique Flex Pressure Nickel Waste water Loss of Internal Surfaces VII.E5.AP-3.3.1-95 C connection boundary aloy (int) material in Miscellaneous 279 A Piping and Ducting Components Flex Pressure Stainless Air-indoor None None IV.E.RP-04 3.1.1-A connection boundary steel (ext) 107 to NRC-15-0068 Page 8 3.2.2.1.8 Miscellaneous ESF Systems in Scope for 10 CFR 54.4(a)(2) Aging Management Programs The following aging management programs manage the effects of aging on nonsafety-related components affecting safety-related systems. Bolting Integrity Buried and Underground Piping

• External Surfaces Monitoring Flow-Accelerated Corrosion Internal Surfaces in Miscellaneous Piping and Ducting Components
• One-Time Inspection Periodic Surveillance and Preventive Maintenance Water Chemistry Control - BWR to NRC-15-0068 Page 9 Table 3.3.1 Summary of Aging Management Programs for the Auxiliary Systems Evaluated in Chapter VII of NUREG-1801 Table 3.3.1: Auxiliary Systems Aging Item Aging Effect/

Management Further Evaluation Number Component Mechanism Programs Recommended Discussion 3.3.1-95 Copper alloy, Loss of material Chapter XI.M38, No Loss of material for most stainless steel, due to pitting, "Inspection of copper alloy, nickel alloy j A nickel alloy, crevice, and Internal Surfaces in and stainless steel steel piping, microbiologically Miscellaneous components exposed to piping influenced Piping and Ducting waste water or components, corrosion Components" condensation is managed and piping by the Internal Surfaces in elements, heat Miscellaneous Piping and exchanger Ducting Components components, Program. The Periodic piping, piping Surveillance and components, Preventive Maintenance and piping Program uses periodic elements; tanks visual inspections or other exposed to NDE techniques to waste water, manage loss of material for condensation other components exposed (internal) to waste water. Steel components exposed to condensation are addressed in Item 3.3.1-89. to NRC-15-0068 Page 10 Table 3.3.2-6 Compressed Air Systems Summary of Aging Management Evaluation Table 3.3.2-6: Compressed Air Systems Aging Effect Aging Component Intended Requiring Management NUREG-Table 1 Type Function Material Environment Management Programs 1801 Item Item Notes Strainer Filtration CGpper Condensation Loss of Compressed Air V3.3AP-15A B a4ey (ext) material Monitoring 24 3.3.1-56 Stainless VII.D.AP-81 steel Strainer Filtration Cepper Condensation Loss of Compressed Air VtD.AP-33.1 A B ally (int) material Monitoring 24G 3.3.1-56 Stainless VII.D.AP-81 B steel t PFeSue Careen r-lndee Leas-of Exer a4 V A11 0 3.3.1 78 A Moiteorng Stainer P-esure Care (ndensation Lees-of Copr escod Air VIL D A-26 3-3--55 B uid ( t4 matea4 M to NRC-15-0068 Page 11 Table 3.3.2-10 Emergency Diesel Generator System Summary of Aging Management Evaluation Table 3.3.2-10: Emergency Diesel Generator System Aging Effect Aging Component Intended Requiring Management NUREG-Table 1 Type Function Material Environment Management Programs 1801 Item Item Notes Tubing Pressure Stainless Raw water Loss of Service Water VII.H2.AP-3.3.1-A boundary steel (int) Material Integrity 55 41 Tubing Pressure Stainless Treated water Loss of Water Chemistry VIl.C2.A-52 3.3.1-D boundary steel fint) material Control - Closed 49 C Treated Water Systems Tubing Pressure Stainless Treated water Cracking Water Chemistry VII.C2.AP-3.3.1-C boundary steel > 140°F (int) Control - Closed 186 43 Treated Water Systems Valve body Pressure Carbon Condensation Loss of Compressed Air VII.D.A-26 3.3.1-D boundary steel int material Monitoring 55 Valve body Pressure Carbon Lube oil (ext) Loss of Oil Analysis VII.H2.AP-3.3.1-A, 302 l boundary steel material 127 97 Valve body Pressure Carbon Lube oil (int) Loss of Oil Analysis VII.H2.AP-3.3.1-A, 302 boundary steel material 127 97 to NRC-15-0068 Page 12 Table 3.3.2-15 Fuel Oil Systems Summary of Aging Management Evaluation Table 3.3.2-15: Fuel Oil Systems Aging Effect Aging Component Intended Requiring Management NUREG-Table 1 Type Function Material Environment Management Programs 1801 Item Item Notes Tank Pressure Carbon Soil (ext) Loss of Aboveground VII.H1.A-3.3.1-A boundary steel Material Metallic Tanks 402 129 Tubing Pressure Carbon Air-indoor Loss of External VIl.L.A-77 3.3.1-78 A boundary steel (ext) material Surfaces Monitoring D Tubing Pressure Carbon Fuel oil (int) Loss of Diesel Fuel Vll.H1.AP-3.3.1-70 A, 303 boundary steel material Monitoring 105 Tubing Pressure Stainless Air-indoor None None VII.J.AP-3.3.1-A boundary steel (ext) 123 120 to NRC-15-0068 Page 13 Table 3.4.2-3-4 Heater Drains System Nonsafety-Related Components Affecting Safety-Related Systems Summary of Aging Management Evaluation Table 3.4.2-3-4: Heater Drains System, Nonsafety-Related Components Affecting Safety-Related Systems Aging Effect Aging Component Intended Requiring Management NUREG-Table 1 Type Function Material Environment Management Programs 1801 Item Item Notes Valve body Pressure Carbon Waste water Loss of material Periodic VII.E5.AP-3.3.1-E boundary steel (int) Surveillance 281 91 and Preventive Maintenance Valve body Pressure Carbon Air - indoor Loss of material External F boundary steel with (ext) Surfaces nickel allov Monitoring clad Valve body Pressure Carbon Treated Cracking - TLAA - metal F boundary steel with water (int) fatigue fatigue nickel alloy clad Valve body Pressure Carbon Treated Loss of material Water F boundary steel with water (int) Chemistry nickel alloy Control - BWR clad Valve body Pressure Copper alloy Air - indoor None None VIll.l.SP-6 3.4.1-A boundary > 15% Zn or (ext) 54 > 8% Al to NRC-15-0068 Page 14 Table 3.4.2-3-6 Condenser and Auxiliaries System Nonsafety-Related Components Affecting Safety-Related Systems Summary of Aging Management Evaluation Table 3.4.2-3-6: Condenser and Auxiliaries System, Nonsafety-Related Components Affecting Safety-Related Systems Aging Effect Aging Component Intended Requiring Management NUREG-Table 1 Type Function Material Environment Management Programs 1801 Item Item Notes Cooler Pressure Carbon Treated water Loss of material Water VIII.E.SP-3.4.1-C, 401 housing boundary steel (int) Chemistry 77 15 Control - BWR Ejector Pressure Carbon Air-indoor Loss of material External VIlI.H.S-29 3.4.1-A boundary steel (ext) Surfaces 34 Monitoring Ejector Pressure Carbon Steam (int) Loss of material Water VIII.A.SP-3.4.1-D, 401 boundary steel Chemistry 71 14 Control - BWR Elector Pressure Carbon Treated water Loss of material Water VIII.E.SP-3.4.1-D, 401 F boundary steel (int) Chemistry 73 14 Control - BWR E-jeeter Presse Staie /\\rndoor Ne Ne 444-4-SP-- - A b aY etee (ex 42 58 Ere P reesm Sntain4ess Trcated wter Loss of material Wate Vl4E-P .41- , 491 aG4ee4: " x'~49-{4 ______be_____ da____ etee___ Chemntro B\\7 4______ to NRC-15-0068 Page 15 4.3 METAL FATIGUE 4.3.1 Class 1 Fatique Analyses Fatigue evaluations were performed in the design of Class 1 components. Class 1 fatigue evaluations are contained in analyses and stress reports, and because they are based on a number of transient cycles assumed for a 40-year operating term, these evaluations are considered TLAA. Based on the numbers of cycles accrued to date, the numbers of cycles at the end of 60 years of operation were projected as shown in Table 4.3-1. The projections are linear projections based of the rate of the occurrence from January 1, 2000, through December 31, 2012, except where identified by footnotes. Fermi 2 recently reviewed the transient cycles that require counting and updated the cycle counts. These reviews provide the basis for the transient cycles that are listed in Table 4.3-1. The Fatigue Monitoring Program (Section B.1.17) tracks these transient cycles to manage the effects of fatigue for Class 1 components. The review included evaluation of transient cycle assumptions for locations that had been exempt from fatigue. The review identified one location locations that were previously exempt from fatigue evaluation meet the fatigue exemption criteria considering 60 years of operation. The Fatigue Monitoring Program tracks transient cycles and requires corrective actions if the numbers of cycles approach analyzed values. The Fatigue Monitoring Program ensures that the numbers of transient cycles experienced by the plant remain within the allowable numbers of cycles. It provides for use of cycle-based fatigue or stress-based fatigue monitoring methods if a component's cumulative usage factor based on cycle counting is projected to exceed 1.0 H after the environmentally assisted fatigue (EAF) calculations are complete. Appendix B, Section B.1.17, provides further details on the Fatigue Monitoring Program. to NRC-15-0068 Page 16 4.3.1.1 Reactor Pressure Vessel As described in UFSAR Section 5.4.6.3.1 and shown in UFSAR Figure 5.4-1, the RPV is a vertical, cylindrical pressure vessel with hemispherical heads of welded construction. The vessel design data are listed in UFSAR Table 5.4-1. The RPV thermal cycles are listed in Table 4.3-1. The RPV is designed, fabricated, tested, inspected, and stamped in accordance with the ASME B&PV Code Section III, 1968, Class 1, up to and including summer 1969 Addenda. Sections 4.3.1.2 and 4.3.1.3 provide additional details on the review of feedwater nozzle and underclad cracking. Table 4.3-2 lists the CUFs for the reactor vessel.Ref. 4-161. I H Fermi 2 will monitor transient cycles using the Fatigue Monitoring Program (Section B.1.17) and assure that action is taken if any of the actual cycles approach their analyzed numbers. As such, the Fatigue Monitoring Program will manage the effects of aging due to fatigue on the reactor pressure vessel in accordance with 10 CFR 54.21(c)(1)(iii). to NRC-15-0068 Page 17 Table 4.3-2 Reactor Pressure Vessel Cumulative Usage Factors General Location Location/Node CUF Nozzle-vessel 0.683 CRD return nozzle intersection Cap Bounded by the nozzle Nozzle-vessel 0.258 intersection Recirculation outlet nozzles Safe end 0.103 Nozzle 0.116 Nozzle-vessel 0.054 intersection Recirculation inlet Safe end 0.002 nozzles Liner, Cut 2, inside 0.220 surface Core AP nozzle Cut II, outside surface 0.637 CRD nozzles Cut 6 0.645 Basin seal skirt 0.162 Shroud support Cut 3a 0.111 4" Vent nozzle bolts 0.318 6" Instrument/head 0.382 spray nozzle bolts CRD assembly Main flange 0.146 H to NRC-15-0068 Page 18 4.3.1.4 Reactor Pressure Vessel Internals As described in UFSAR Section 4.1.2, the major Fermi 2 reactor internal components are the core (fuel, channels, control rods, and instrumentation), the core support structure (including the core shroud, shroud head separators, top guide, and core support plate), the steam dryer assembly, and the jet pumps. Table 4.3-3 identifies the usage factors that were calculated for the RVI locations that are subject to aging management review (ef. 4-16). l H Table 4.3-3 RPV Internals Cumulative Usage Factors General Location Location/Node CUF Core spray lines 0.287 Jet pump riser braces 0.266 Access hole covers Ring to cover 0.380 Ring to adaptor ring 0.163 Adaptor ring to shroud support 0.876 plate Jet pump auxiliary spring Spring half 0.000 wedges Right half 0.000 Feedwater sparger Sparger pipe to spray nozzle 0.598 Thermal sleeve to reducer tee 0.497 H Sparqer pipe to end plate 0.400 to NRC-15-0068 Page 19 4.3.1.5 Reactor Pressure Vessel Internals NPVC 1 (NPVC - 1, 2, 3 Draft ASME Code for Pumps and Valves for Nuclear Power, Class I, 11, Ill). As identified in Note z of UFSAR Table 3.2-1 and UFSAR Table 3.2-4 Note j, the reactor recirculation pumps were upgraded to the 4th generation design, and the modified components were designed and manufactured to ASME III, 1989. Representative analyses of recirculation pumps are summarized in UFSAR Table 3.9-20. The transients that require tracking are included in Table 4.3-1. Table 4.3-4 provides the CUFs for the reactor recirculation pumps Ref. 4-16l H 4.3.1.6 Class 1 Piping Detailed fatigue analyses were generated to analyze multiple locations on each system within the Class 1 boundary. The transients that require tracking are included in Table 4.3-1. Table 4.3-6 provides the highest piping cumulative usage factors (Ref. 4-16), and Table 4.3-7 provides the associated valve cumulative usage factors (Ref. 4-16). H to NRC-15-0068 Page 20 Table 4.3-6 Piping Cumulative Usage Factors General Location Location/Node CUF Div. 2 core spray piping inside containment (CS-01) Node 95 0.035 Div. 1 core spray piping inside containment (CS-02) Node 90 0.032 Div. 2 core spray piping outside containment (CS-03) Node 7 0.007 Div. 1 core spray piping outside containment (CS-05) Node 6 0.011 Main steam piping outside containment (MS-05) Node 960 0.003 RWCU supply piping inside drywell Node 030B 0.043 RWCU supply piping outside drywell Node 387 0.036 Div. 1 & 2 RHR return piping outside drywell (RHR-02) Node 5 0.136 Div. 1 & 2 RHR supply piping outside drywell (RHR-07) Node 9 0.055 Feedwater Loop B piping inside drywell (FW-01) Node 40 0.071 Feedwater Loop A piping inside drywell (FW-02) Node 40 0.101 Feedwater Loop B piping outside drywell (FW-04) Node 16 H Feedwater Loop A piping outside drywell (FW-05) Node 305 RCIC steam supply piping outside drywell (RCIC-01) Node 6 0.001 HPCI steam supply piping outside drywell (HPCI-02) Nodes 6 & 8 0.002 SLC piping inside containment Node 55 0.097H SLC piping outside containment Node 95 0.056 Main steam drain piping inside drywell Node 50 0.047 Main steam drain piping outside drywell Node 50 0.022 to NRC-15-0068 Page 21 Table 4.3-6 Piping Cumulative Usage Factors General Location Location/Node CUF 0.181 RPV vent line from RPV to bulkhead piping Node 17 0.026 RPV vent line from bulkhead to drain manifold piping Node 33W/412 0.026 RPV vent line from bulkhead to main steam A piping Node 43W/397 0228 Main steam line A and HPCI steam piping Node 102 0.085 Main steam line B and RCIC steam piping Node 200 0.174 Main steam line C piping Node 200 0.147 Main steam line D piping Node 200 0.054 Recirculation A and Div. i RHRR piping Node 250 0.012 Recirculation B, RHRS and Div. 2 RHRR piping Node 516 0.011 Condensinq chambers B2102DC06A& 0.410 C Notc 1: The prcjccted 40 year CUF for this location is 0.803. The projected 60 year CUF for this location is greater than 1.0, and this has been identified in the Corrective H Action Program. Potential solutions include repair of the component, replacement of the component, a more rigorous analysis of the component, or rnonitoring and trakig yces o nsrefaigu lmis re otexeeed to NRC-15-0068 Page 22 4.3.3 Effects of Reactor Water Environment on Fatigue Life Industry test data indicate that certain environmental factors (such as temperature and dissolved oxygen content) in the primary systems of light water reactors could result in greater susceptibility to fatigue than would be predicted by fatigue analyses based on the ASME Section Ill design fatigue curves. The ASME design fatigue curves were based on laboratory tests in air at low temperatures. Although the fatigue curves derived from laboratory tests were adjusted to account for effects such as data scatter, size, and surface finish, these adjustments may not be sufficient to account for actual reactor water operating environments. As reported in SECY-95-245, the NRC believes that no immediate staff or licensee action is necessary to deal with the environmentally assisted fatigue issue. In addition, the staff concluded that it could not justify requiring a backfit of the environmental fatigue data to operating plants. However, the NRC concluded that environmentally assisted fatigue should be evaluated for any proposed extended period of operation for license renewal because metal fatigue effects increase with service life. NUREG/CR-6260 addresses the application of environmental correction factors to fatigue analyses (CUFs) and identifies locations of interest for consideration of environmental effects (Ref. 4-13). NUREG/CR-6260 identified the following component locations to be the most sensitive to environmental effects for General Electric plants. (1) Reactor vessel shell and lower head (2) Reactor vessel feedwater nozzle (3) Reactor recirculation piping (including inlet and outlet nozzles) (4) Core spray line reactor vessel nozzles and associated Class 1 piping (5) Residual heat removal nozzles and associated Class 1 piping (6) Feedwater line Class 1 piping guidance (formulas) providod in NUREG/CR 6909 (Ref. 1 15) to calculate environmontally assisted fatigue correction factors (F)E for nickl alloy components but specifies that NUREG/CR 6583 (Ref. I 11) may be uscd for carbon and low alloy sto!, and NUREG/CR 5704 (Ref. A 12) may be us ed for astcnitic stainless stool Environmentally assisted fatigue (EAF) screening was performed (Ref. 4-17) for these H NUREG/CR-6260 locations and the remaining ASME Class 1 reactor pressure vessel and piping locations for which fatigue had been assessed that are (1) wetted (in contact with liquid reactor coolant) and (2) form part of the reactor coolant pressure boundary. All wetted reactor coolant pressure boundary locations with CUF calculations were addressed, so determination of thermal zones was not required. The components with the highest calculated CUF in each svstem were evaluated for the effects of EAF. to NRC-15-0068 Page 23 The evaluation of environmentally assisted fatigue included an evaluation of the water chemistry history to determine the cumulative environment for the components when determining the dissolved oxygen (DO) Values were determined for six zones in the primary system, using data from the three chemistry regimes used historically (Normal Water Chemistry, Hydrogen Water Chemistry, and On-line Noble Metal Chemistry). Using the availability data, there was the equivalent of over 12 fuli years of hydrogen water chemistry (HWC) from 1997 to the end of 2012. A 95 percent availability in 2013 through March 2015 would resuit in over 30 moro years of HWC.This corconlng analysis will therofore use 12 yoars as a best cotimate of thc tim~e for HWC (42 yars with HWC and 18 [60 42] years without HWC). implementation of HWC, the oxygen consentrations in the vessel and attached piping (othcr tha nf foodater) aro loss than 5 ppb. A representative v:alue for feedwater following HC i 35 ppb. The screening evaluation used guidance in NUREG/CR-6909 (Ref. 4-15) as allowed by NUREG-1801, Revision 2. The fatigue curves in NUREG/CR-6909 were applied for all materials, The methodology discussed in EPRI 1024995, "Environmentally Assisted Fatique Screening: Process and Technical Basis for Identifying EAF Limiting Locations" was used as guidance to determine which locations are bounding. These locations are referred to as Sentinel locations. The Sentinel locations are monitored to manage the effects of aging including EAF in the period of extended operation. The environmental correction factor (Fe,) calculation followed the formulas in NUREG/CR-6909. The historical dissolved oxygen chemistry input from the plant was used to account for the various chemistry zones and regimes that existed in the plant operating history and future projections when calculating the correction factor (Fe1) for 60 years. Fatigue calculations for 60 years used the projected cycles from Table 4.3-1. The results of the EAF screening calculation were used to determine the bounding components based on fatigue, including EAF effects. To summarize the approach taken in the EAF screening evaluation, the following steps were followed for each one of the components that have calculated CUFs: Pre-screening to establish the eligibility of the component for further EAF evaluation (e.g. wetted or not). Determine the bounding location in terms of fatigue for the component. Apply NUREG/CR-6909 fatigue curves. Obtain alternating stresses from design information.

Enclosure I to NRC-15-0068 Page 24 Calculate the environmental correction factor, Fen, using the formulas in NUREG/CR-6909 and the least favorable conditions for sulfur content in metal strain rate and temperature in thermal transient along with the zone-specific DO values. Apply time weighted average Fen factor to increase the fatigue usage. Compare against EAF criterion of 1.0 and the Sentinel location screening threshold value of 0.8. Thc following equations were utilized in determining environmental fatigue correction factors. Carbon Stoe! The environrentally asisted fatigue correction factor (F.) for carbon steel is calculated using NUREG/CR 6583, Equation 6.5a. Lo A!! I v Stee! The environrmentely assisted fatigue correction factor (F ) for low alloy steel is calculated using NUREG/CR 6583, Eq. 6.5b. The environmentally assisted fatigue correction factor (F) for wrought and east The environrmentally assisted fatigue correction factor (F ) for Ni Cr Fe alleys is calculated using NUREG/CR 6909, Eq A.14. To recalculate F using NUREG 6909 requires the use of updated fatigue tables. This will be addressed by the cernritment identified in this section to complete a reanalysis prior to the period of extended operation. The screening calculations determined that some components have a calculated CUFen greater than 1.0. Where this occurred, the locations were re-evaluated with reduced Fen multipliers using average transient temperatures (based on NUREG/CR-6909 guidance) or average load pair temperatures where available. The results of the EAF screening are shown in Table 4.3-8. Sentinel locations are noted in Table 4.3-8. As shown in Table 4.3-8, this screening has determined that there are locations that, when accounting for environmental effects, have projected usage factors greater than 1.0. Additional action will be needed, e.g. more detailed analysis or stress-based or cycle-based fatigue monitoring, as part of the Fatigue Monitoring Program (Section B.1.17) for these locations. DTE H will update the fatigue usage calculations using refined fatigue analyses to determine va!!d CUFs less than 1.0 when accounting for the effects of reactor water environment prior to the period of extended operation or institute stress-based fatigue (SBF) or cycle-based fatigue to NRC-15-0068 Page 25 (CBF) monitoring to demonstrate CUFs remain less than 1.0. This includes applying the annennrieo C factore to valird CUN dcelterminedi useinga n ADC approve verson of the ASCMEl cae or ARC anprover alternative (o NAD anproed oe ase). DTwi reviw dsign basis ASMAAC laos i omponent fatirgue cvahuatinn to ensrr theem locatinseauhed ompnnento within the roector oolant nep re neouncary E-nvironmental effccts on fa+teguc for these critisal onmnonents will be alu-h ated sing one of the following setsof formLa. narIhfln fnlfC Ino l lem nII 4*,, 10p Thoen peroi in AllEIDCR O

656, sing the applicable ASMEA Sction I fatigre e igin crv e.

Thoe previrderd in Appeondix A of AlIDC/DR

909, using either the appicabl ASM Sction Ill fatigce sgnr cre o rr the fatigue esign curnve for carbn and lew anoy steel provided in NIUREGDCR 6909 (Figures A. 4 and A.2, respectively, T

n H -An ARC aoedor alternative. /\\ e onitio eiloncn l Theos perided in A IDC/CR 570A, using the enlichn ASME SectonieR I Thoen prvdedl in AeG iCR/NDRE 6 ingr the fatirge oenr crv for astfaniti stainless steel pevided in NREC/CR 6909 (Figure A3 nd Table A2). . An ADC aonea lternative. AlTholoserovie nNRG 99 sn h aiu eincrefraseii stainless steel providedr in All IDC/DROO 69 ir9 (ire A. 3 anrd Table A2). .A n ADC approved alternative. Fermi 2 manages the effects of fatigue, including environmentally assisted fatigue, under the Fatigue Monitoring Program (Section B.1.17) for the period of extended operation in accordance with 10 CFR 54.21(c)(1)(iii). to NRC-15-0068 Page 26 Table 4.3-8 EAF Screening of Fermi 2 Locations NUREG/CR-6260 Fermi 2 Generic Location Location Material 1" CUF Fen EAF CUF 1 Reactor vessel shell Reactor vessel LAS 0.021 7.21 0.153 and lower head shell? 0-954 5-7-2 0.399 1 Reactor vessel shell CRD nozzle? NBAM 0.630 1 - 1.284 and lower head 9.045 3.453 >4 2 Reactor vessel FW nozzle safe CS 0.087 1.88 0.164 plus feedwater nozzle end (CS 0.257 4-74 0.0014 rapid portion)? cycling = 0.165 0-465 2 Reactor vessel FW nozzle safe SS 0.664 6.6_- 6.364 plus feedwater nozzle end (SS 0-585 983 0.0014 rapid portion)? 45.35 cycling = 6.374 >4 2 Reactor vessel Nozzle-vessel LAS 0.0154 6.10 0.0942 0H1-feedwater nozzle intersection? 0437 5-2 plus 0.0206 0-924 rapid cycling = 0.115 0-233 3 Reactor recirculation RR inlet nozzle SS 0-20 43.25 Note 5 piping (including inlet liner >4 and outlet nozzles) 3 Reactor recirculation RR inlet nozzle SS 0.006 11.20 0.071 piping (including inlet safe end? 9092 13.25 9-927 and outlet nozzles) 3 Reactor recirculation RR inlet nozzle LAS 0.023 4.06 0.095 piping (including inlet nozzle-vessel 0-054 5-2 0-39 and outlet nozzles) intersection? 3 Reactor recirculation RR outlet LAS 0.112 4.06 0.453 piping (including inlet nozzle nozzle-0-25 5-2 >4 and outlet nozzles) vessel intersection? to NRC-15-0068 Page 27 Table 4.3-8 EAF Screening of Fermi 2 Locations NUREG/CR-6260 Fermi 2 Generic Location Location MaterialW CUF Fen EAF CUF 3 Reactor recirculation RR outlet SS 0.021 11.2 0.240 piping (including inlet nozzle safe 0.103 13.25 >4 and outlet nozzles) end2 RR outlet LAS 0.11- &-72 Bounded by nozzle nozzle-vessel intersection 07084 3 Reactor recirculation RR piping-SS 0.018 11.2 0.199 piping (including inlet 0 13.25 0.159 and outlet nozzles) 3 Reactor recirculation RR valve SS 0.215 43.25 Bounded by piping (including inlet adiacent piping and outlet nozzles) >4 4 Core spray line reactor Core spray LAS 0.048 5.31 0.254 vessel nozzle and nozzle -vessel 0-144 5-7-2 0-52 H associated Class 1 intersectionz piping 4 Core spray line reactor Core spray NBAM 0.075 3.65 0.273 vessel nozzle and nozzle safe 0.079 348 0.251 associated Class 1 end2 piping 4 Core spray line reactor Core spray CS 0.0152 4.95 0.075 vessel nozzle and valve 0.048 4-774 07084 associated Class 1 piping 4 Core spray line reactor Core spray CS 0.0091 4.95 0.045 vessel nozzle and piping? 0.035 444 0.051 associated Class 1 piping 5 Residual heat removal RHR valve CS 0.063 3.78 0.239 (RHR) nozzles and 0440 4-97 0.570 associated Class 1 piping to NRC-15-0068 Page 28 Table 4.3-8 EAF Screening of Fermi 2 Locations NUREG/CR-6260 Fermi 2 Generic Location Location Material 1 CUF Fen EAF CUF 5 Residual heat removal RHR piping? CS 0.048 3.78 0.180 (RHR) nozzles and .-436 4, Q-554 associated Class 1 piping 6 Feedwater line Class 1 FW valve CS 0.063 3.78 0.239 piping 40 7-4 0-244 6 Feedwater line Class 1 FW piping-CS 0.0003 3.78 0.0012 piping 0-246 44 0-428 RWCU piping CS 0.014 3.78 0.052 RWCU valve CS 0.018 3.78 0.069 HPCI valve CS 0.053 1.88 0.099 SLCS piping SS 0.108 2.42_- 0.503 9.733 H Containment penetrations: X-13A/B, RHR CS 0.086 3.78 0.326 return weld 2 Highest non-CS 0.181 3.78 0.685 Sentinel: X-12 body, RHR supply Condensing SS 0.385 9.73 3.754 chamber 2 Core AP NBA 0.580 2.55 1.484 nozzle 2 CRD assembly, SS 0.134 2.45 0.329 main flange to NRC-15-0068 Page 29 Table 4.3-8 EAF Screening of Fermi 2 Locations NUREG/CR-6260 Fermi 2 Generic Location Location Materiall CUF Fen EAF CUF RR pump SS 0.150 3.86 0.581 cooler2

1. CS: carbon steel. LAS: low alloy steel. NBA: nickel-based alloy. SS: stainless steel.

2. This component is a sentinel location.

3. Fen was calculated for each transient or load pair based on average temperature.

4. CUFen is above 1 so additional action will be needed, e.g. more detailed analysis or stress H

based monitoring.

5. Not part of reactor coolant pressure boundary, so no EAF evaluation is needed.

.CS:cabon otool. L.A.? low alloy stcel. NBA: nickel bascd alley. SS: ctainlcG S'tool

b. This is a nickel alloy Iccatian that will requir~e roanalysis of usage factor with new fatigue to NRC-15-0068 Page 30 4.6 CONTAINMENT LINER PLATE, METAL CONTAINMENT, AND PENETRATIONS FATIGUE ANALYSES 4.6.1 Primary Containment The usage factors are identified in Table 4.6-1 (Ref. 4-16). The SRV actuations and seismic l H cycles are tracked and will be maintained below the cycle value used in the fatigue evaluation, or reanalysis will be completed. Fermi 2 will manage the aging effects due to fatigue using the Fatigue Monitoring Program (Section B.1.17) in accordance with 10 CFR 54.21(c)(1)(iii).

4.6.5 Containment Penetrations As described in UFSAR Section 3.8.2.1.3.1, sleeved penetration assemblies with bellows consist of the process pipe, guard pipe, penetration sleeve bellows, and flued head. For Class 1 piping (Fermi 2 Group A), the design of the flued head meets ASME Section III Class 1 requirements, which specify a fatigue analysis that determines the cumulative usage factor for the flued head. UFSAR Figure 3.8-9 provides a cross-sectional drawing of the penetration assemblies and a listing of the penetrations that utilize this design (designated as Type 1). The usage factors are shown in Table 4.6-2 for these flued head penetrations (Ref. 4-16) based on l H the number of cycles shown in the analysis input value column in Table 4.3-1. to NRC-15-0068 Page 31 Table 4.6-2 Cumulative Usage Factors for Flued Head Penetrations Location/Node CUF Penetrations X-9A / B (Feedwater A, B) 0-464 0.471 H Penetration X-10 (RCIC steam supply) 0.046 Penetration X-12 (RHR supply) 0.416 Penetrations X-1 6A / B (Core spray A, B) 0.405 Penetrations X-7A-D (Main steam lines A 0.271 through D) Penetration X-8 (Main steam line drains) 0.094 Penetration X-11 (HPCI steam supply) 0.020 Penetrations X-13A / B (RHR return A, B) 0.371 Penetration X-43 (Reactor water clean up) 0.-267 0.144 H Penetration X-42 (Standby liquid control) 0.002 to NRC-15-0068 Page 32 4.7 OTHER PLANT-SPECIFIC TLAAS 4.7.2 Determination of High-Energy Line Break Locations UFSAR Sections 3.6.1 and 3.6.2 state that the method used to determine the intermediate locations of pipe breaks in high-energy lines includes an evaluation based on CUFs being less than 0.1 if other stress criteria are also met. Design criteria for piping between the primary containment and outboard isolation valves provide for maximum stresses considering all normal and upset conditions as calculated by the equations in Paragraph NB-3653 of ASME B&PV Code Section III. UFSAR Section 3.6.2.1.2.2 states that pipe breaks were not postulated in the high energy piping between the containment penetration and outboard isolation valves since the piping was conservatively designed and restrained. The calculated CUFs for containment penetration piping were also limited to values less than 0.1 if equation 10 of NB-3653 exceeds 2.4 Sm. The CUFs, as calculated in the design fatigue analyses, are based on the design transients assumed for the original 40-year life of the plant; therefore, the CUF analyses used in the selection of postulated high-energy line break locations are considered TLAAs. The Fatigue Monitoring Program (Section B.1.17) identifies when the transients affecting high-energy piping systems are approaching their analyzed numbers of cycles. The mod fiation of the Fcrmi 2 fatigue calculations to accounRt for the projebod cycles farF 60 years resultod in completd as ncessary. Re-analyses of the limiting locations where a break is not already assumed for 60 years of projected cycles resulted in CUFs meeting the 0.1 criterion for high-energy line break exclusion (Ref. 4-16). DTE Electric will manage the effects of aging associated with the fatigue analyses used in the selection of postulated high-energy line break locations using the Fatigue Monitoring Program (Section B.1.17) in accordance with 10 CFR 54.21(c)(1)(iii). to NRC-15-0068 Page 33

4.8 REFERENCES

4-12 NUREG/CR 570 (\\NL 98/31), Effcta of LWR Coo/ant Environmonta on FatiguH Dosign Cartes~ of Austonit/o Stainlona Stools, AprIl 1999. Deleted 4-13 N U REG/CR-6260, (IN EL 95/0045) Application of NUREG/CR-5999 Interim Fatigue Curves to Selected Nuclear Power Plant Components, February 1995. 4-14 NUREG/CR 6583 (ANL 97/18), Effects of LWR Coolant Environmonta en Fatgue H Design Cua of Carbon and Low !ey Steel, Fobruary 1998. Deleted 4-15 NUREG/CR-6909 (ANL-06/08), Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials, February 2007. 4-16 Design Calculation DC-6222, "ASME Operating Plant Fatigue Assessment for License Renewal - RCPB Components", Revision A, March 2015. H 4-17 Structural Integrity Associates Calculation 1400966.301, "Fermi 2 Environmental Assisted Fatigue (EAF) Screening", Revision 2, May 2015. to NRC-15-0068 Page 34 A.1.43 Water Chemistry Control - BWR The Water Chemistry Control - BWR Program manages loss of material, cracking, and fouling in components exposed to a treated water environment through periodic monitoring and control of water chemistry. The Water Chemistry Control - BWR Program monitors and controls water chemistry parameters such as pH, chloride, conductivity, and sulfate. EPRI Report 1016579 3002002623 is used to provide guidance. The One-Time Inspection Program utilizes inspections or non-destructive evaluations of representative samples to verify that the Water Chemistry Control - BWR Program has been effective at managing aging effects. The representative sample includes low flow and stagnant areas. to NRC-15-0068 Page 35 A.2.2 Metal Fatigue A.2.2.1 Class 1 Metal Fatique Analyses Fatigue evaluations were performed in the design of the Fermi 2 Class 1 components. Class 1 fatigue evaluations are contained in analyses and stress reports, and because they are based on a number of transient cycles assumed for a 40-year operating term, these evaluations are considered time-limited aging analyses. The Fatigue Monitoring Program (Section A.1.17) tracks transient cycles and requires corrective actions if the numbers of cycles approach analyzed values. It provides for use of cycle-based fatigue or stress-based fatigue monitoring methods if a component's cumulative usage factor H based on cycle counting is projected to exceed 1.0 after the environmentally assisted fatique (EAF) calculations are complete. The Fatigue Monitoring Program will manage the effects of aging due to fatigue in accordance with 10 CFR 54.21(c)(1)(iii). The following provides additional information for specific Class 1 components. A.2.2.3 Effects of Reactor Water Environment on Fatigue Life NUREG/CR-6260 addresses the application of environmental correction factors to fatigue analyses (cumulative usage factors [CUFs]) and identifies locations of interest for consideration of environmental effects. NUREG/CR-6260 identified the following component locations to be the most sensitive to environmental effects for General Electric plants. (1) Reactor vessel shell and lower head (2) Reactor vessel feedwater nozzle (3) Reactor recirculation piping (including inlet and outlet nozzles) (4) Core spray line reactor vessel nozzles and associated Class 1 piping (5) Residual heat removal nozzles and associated Class 1 piping (6) Feedwater line Class 1 piping Environmentally assisted fatique (EAF) screening was performed for these NUREG/CR-6260 locations and the remaining ASME Class 1 reactor pressure vessel and piping locations for which fatigue had been assessed that are (1) wetted (in contact with liguid reactor coolant) and (2) form part of the reactor coolant pressure boundary. The components with the highest calculated CUF in each system were evaluated for the effects of EAF. The screening evaluation H used guidance in NUREG/CR-6909, as allowed by NUREG-1801, Revision 2. The fatigue curves in NUREG/CR-6909 were applied for all materials. The methodology discussed in EPRI 1024995, "Environmentally Assisted Fatigue Screening: Process and Technical Basis for Identifying EAF Limiting Locations" was used as guidance to determine which locations are bounding These locations are referred to as Sentinel locations. The Sentinel locations are to NRC-15-0068 Page 36 monitored to manage the effects of aging including EAF in the period of extended operation. Fermi 2 perforrned a scning evaluation of these six locations using the guidance provided in NUREG 1801, Revisien 2. This screening has determined there are locations that, when accounting for environmental effects, have projected usage factors greater than 1.0. Additional action will be needed, e.g. more detailed analysis or stress-based or cycle-based fatigue monitoring, as part of the Fatigue Monitoring Program (Section A.1.17) for these locations. Fermi 2 will update the fatigue usage calculatinns using refinred fatigue anryses to determinn the periord of extenedr operation. This inclues applying the appronriae Ce factors to valird CUn rdeterminord sing an ADC aoedr version of the ASMlnl cerde or ANDC aovedsar alternative (e.g., NRC approved cede casc). Fermi 2 will review design basic ASME Class 1 cornponent fatigue evaluations to ensure the Fermi 2 locations evaluated for the effects of the reactor coolant environment en fatigue include the rnost lniting cornponents within the reactor coolant pressure boundary. Environmental effects en fatigue for these critical components wil be eau ted sing one of the following sets of formu lae: SThoep meo idedr in Al URDC!/C 6583 u sing the applicble ASME~ Scartion liI f tigue H design curvye a Those providedr in Appndorix A ef AURGDC/DRO6909, sing either the applicablo alloy steel provided in NUREGICR 600 (Figures A anrd A.2, respectively, and Tahle Ad \\ a An ANDC approverd alternativeo Austenitis stainless steels a Thorse provirderd in All IDC/O 5-70A1, sing the appliable ASCME Section Ill fatirgue design curve. & Thoe providedr in AllEGDC/P

600A, sing the fatigue desrigin crvero for astefnitire stainess steel proided in N IREC/R 909 ( igure A.

anRd Table A.2). a A n ANDC aovredr alternative N-iikel alloes SThose previrdod in AllEIDC/DRO 699 singe the fatige dsigin cumre for as tonitic stainless Steel perovided in All IRC /CR 690 (Figurr A.3 nd Table A 2) a An ADC anorrovd altrenative Fermi 2 will manage the effects of fatigue, including environmentally assisted fatigue, under the Fatigue Monitoring Program (Section A.1.17) for the period of extended operation in accordance with 10 CFR 54.21(c)(1)(iii). to NRC-i5-0068 Page 37 A.3 REFERENCES A.3-1 [Fermi 2 Licce Renewa! Application later] DTE Electric Company to NRC, "Fermi 2 License Renewal Application," NRC-14-0028, letter dated April 24, 2014 (ML14121A554). A.3-2 [NRC Safety Evaluation Report for Fermi 2 License Renewal-later] A.3-3 DTE Electric Company to NRC, "License Amendment Request for Measurement Uncertainty Recapture (MUR) Power Uprate," NRC-13-0004, letter dated February 7, 2013 (ML13043A659). A.3-4 NRC to DTE Electric, "Fermi 2-Issuance of Amendment re: Measurement Uncertainty Recapture Power Uprate (TAC NO. MF0650)," letter dated February 10, 2014 (ML13364A131). to NRC-15-0068 Page 38 A.4 LICENSE RENEWAL COMMITMENT LIST No. Program or Activity Commitment Implementation Source Schedule 14 Fire Water System Enhance Structures Monitoring Program as follows: Prior to .1.19 September 20, S_q. If the decreasing trend in fire water system flow tests is 2024, or the end I K not resolved through the Corrective Action Program of the last prior to the period of extended operation, revise Fire refueling outage Water System Program procedures to continue prior to March 20, performing annual fire water system flow tests during 2025, whichever the period of extended operation until such a time as is later, with the trend data from fire water system flow tests indicates exception that the that the system will be capable of performing its activities intended function throughout the period of extended described in this operation and therefore TRM frequency may be commitment for resumed. piping segments q r. Revise Fire Water System Program procedures to designed to be j K include formal documentation of the CCHVAC makeup dry but and recirculation fire water supply drain down determined to be inspection for indications of flow blockage. collecting water shall be conducted within five years prior to March 20, 2025. to NRC-15-0068 Page 39 Table B-3 Fermi 2 Program Consistency with NUREG-1801 NUREG-1801 Comparison Consistent with Programs with NUREG-Programs with Exception to NUREG-Plant-Program Name 1801 Enhancement 1801 Specific Water Chemistry X X Control - BWR to NRC-15-0068 Page 40 B.1.17 FATIGUE MONITORING Enhancements Element Affected Enhancement

1. Scope of Program Revise Fatigue Monitoring Program procedures

2. Preventive Actions so that the scope of the program includes

3. Parameters Monitored or monitoring the operating hours for the main Inspected steam bypass operation at the 30%-45% valve

4. Detection of Aging Effects open position and perform trending to ensure

5. Monitoring and Trending that the operating time for the main steam

6. Acceptance Criteria bypass operation remains below the design limit

7. Corrective Actions during the period of extended operation.

Revise Fatigue Monitoring Program procedures to provide for correctivee nctins to p revetr the operating tirne for the rnain steam bypass frorn eceeding the analysis during the period of extended operation. cceptable corrective actins inclu de repair of the compeoent, replacement of the component, or a more rigorous analysis of the cornponent to demnstr--te that the enviGe life will not b exceeded during the period of extended _____operation.___L

7. Corrective Actions Revise Fatigue Monitoring Program procedures to provide for corrective actions to prevent the operating time for the main steam bypass from exceeding the analysis during the period of extended operation. Acceptable corrective actions include repair of the component, replacement of the component, or a more rigorous analysis of the component to demonstrate that the service life will not be exceeded during the period of extended operation.

to NRC-15-0068 Page 41 B.1.43 WATER CHEMISTRY CONTROL - BWR Program Description The Water Chemistry Control - BWR Program manages loss of material, cracking, and fouling in components exposed to a treated water environment through periodic monitoring and control of water chemistry. The Water Chemistry Control - BWR Program monitors and controls water chemistry parameters such as pH, chloride, conductivity, and sulfate. EPRI Report 1016579 3002002623 is used to provide guidance The One-Time Inspection Program utilizes inspections or nondestructive evaluations of representative samples to verify that the Water Chemistry Control - BWR Program has been effective at managing aging effects. The representative sample includes low flow and stagnant areas. NUREG-1801 Consistency The Water Chemistry Control - BWR Program is consistent with the program described in NUREG-1801, Section XI.M2, Water Chemistry, with one exception. Exceptions to NUREG-1801 The Water Chemistry Control - BWR Program has the following exception. Element Affected Exception N/A (Program Description) NUREG-1801, Section XI.M2, Water Chemistry, references the guidelines in BWRVIP-190, EPRI 1016579 as the basis of the water chemistry program for boiling water reactors under Program Description. Fermi 2 has adopted Rev. 1 of BWRVIP-190, BWR Water Chemistry Guidelines, EPRI 3002002623.1 Exception Note

1. The BWR Water Chemistry Guidelines were updated in 2014 to provide an enhanced methodology for establishing site-specific BWR water chemistry control programs. The focus is on mitigating environmentally assisted cracking, maintaining fuel performance, controlling flow accelerated corrosion and controlling radiation fields. The guidelines were separated into two volumes - the first on implementation guidance and the second on the technical basis and background information. EPRI reports such as "BWR Water Chemistry Guidelines" are industry reports, which are reviewed and revised by industry experts to incorporate recent industry operating experience to NRC-15-0068 Page 42 and best practices. Adoption of the revised guidelines is required.

Through use of the updated BWR Water Chemistry Guidelines, which are based on operating experience with the goal to improve performance, the Fermi 2 Water Chemistry Control - BWR Program will better manage the effects of aging on materials exposed to the BWR treated water environment. Key non-editorial changes in the guidelines include the following: A new section on water chemistry guidance for fuel reliability was added. Guidance on monitoring and eliminating transient sources of ionic impurities in reactor water to minimize impacts on accelerated Intergranular Stress Corrosion Cracking (IGSCC) and potential adverse effects on fuel was included. Guidance was refined for feedwater iron, feedwater copper and feedwater zinc action levels and needed values for fuel performance. Good practice recommendations for chemistry control in auxiliary systems, with emphasis on minimizing impurities that can potentially enter the reactor coolant were added. A new chapter on startup good practices and conditions to avoid based on operating experience was added. Recommendations were enhanced on shutdown good practices and conditions to avoid based on experience, including mid-cycle outages in addition to refueling outages. Tables on data monitoring and evaluation, including good practice sampling frequencies, were updated, and half-lives for activated corrosion products and fission products were included. The technical basis for water chemistry control in BWRs was consolidated and updated, including addition of information on zinc oxide and depleted zinc oxide. The bases for environmentally assisted fatigue of structural materials, low-alloy steel cracking, the effects of impurities on crack growth rates, stress corrosion cracking of carbon steel, and IGSCC mitigation methods were updated. Latest research results were added on controlling dissolved oxygen to minimize flow accelerated corrosion and the effects of noble metal chemistry and on-line noble metal chemistry. Water chemistry impacts on fuel integrity were updated with experience, including an additional table of chemistry trigger values for fuel risk assessment. Latest industry experience was used to update the chapter on water chemistry effects on radiation fields, including effects of moderate hydrogen water chemistry, noble metal chemistry addition, on-line noble chemistry, zinc iniection and operating with low iron. A section was added on good practices and conditions to avoid for the to NRC-15-0068 Page 43 design and operation of condensate polishing systems in order to optimize feedwater quality, with emphasis on minimizing the potential for impurities that may impact fuel performance. Sampling practices and monitoring techniques were updated, including background information, experience and good practices for electrochemical corrosion potential monitoring and on monitoring moisture carryover. The guidance on developing a BWR strategic water chemistry plan was updated. The tables of BWR transients have been updated and the latest results of higher conductivity crack growth rate studies have been incorporated. The correction in measured conductivity for presence of ionic species that are benign towards system integrity and to bound concentrations of potential ionic impurities that are not directly measured were updated. Improvements in selecting the power profile, gamma and fast neutron dose rates and chemical reaction rate constants and radiolysis G-values were added. A new appendix was added requiring all plants that inject hydrogen for IGSCC mitigation to have a mitigation performance indicator and providing good practices. Information on ultrasonic fuel cleaning was removed from this document and placed in a different EPRI guideline on fuel reliability. Another appendix was added on use of new technologies, such as methanol and titanium dioxide. Appendices were added for conversion to SI units. All changes were conservative, or provided additional information. Enhancements None}}